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A first-of-its-kind, inductively driven micro-particle (Pellet) accelerator and injector have been developed and operated successfully in ADITYA-U circular plasma operations, which may ably address the critical need for a suitable disruption control mechanism in ITER and future tokamak. The device combines the principles of electromagnetic induction, pulse power technology, impact, and fracture dynamics. It is designed to operate in a variety of environments, including atmospheric pressure and ultra-high vacuum. It can also accommodate a wide range of pellet quantities, sizes, and materials and can adjust the pellets’ velocities over a coarse and fine range. The device has a modular design such that the maximum velocity can be increased by increasing the number of modules. A cluster of lithium titanate/carbonate (Li _2 TiO _3 /Li _2 CO _3 ) impurity particles with variable particle sizes, weighing ∼50–200 mg are injected with velocities of the order of ∼200 m s ^−1 during the current plateau in ADITYA-U tokamak. This leads to a complete collapse of the plasma current within ∼5–6 ms of triggering the injector. The current quench time is dependent on the amount of impurity injected as well as the compound, with Li _2 TiO _3 injection causing a faster current quench than Li _2 CO _3 injection, as more power is radiated in the case of Li _2 TiO _3 . The increase in radiation due to the macro-particle injection starts in the plasma core, while the soft x-ray emission indicates that the entire plasma core collapses at once.

The ADITYA upgrade (ADITYA-U), a medium-sized $\left( {{R_0} = 75{\text{ cm}},\,\,a = 25{\text{ cm}}} \right)$ conventional tokamak facility in India, has been consistently producing experiments findings by using circular and shaped-plasmas. Recognizing the plasma parameters aligning closely with the design parameters of circular limited plasmas, ADITYA-U shifted its focus toward exploring the operational regime for experimentation on saw-tooth and MHD phenomena. Moreover, ADITYA-U has made consistent advancements toward conducting preliminary plasma shaping experiments through the activation of top and bottom divertor coils utilizing hydrogen as well as deuterium fuels. Confinement is improved by a factor of ∼1.5 in ${D_2}$ plasmas when compared to H _2 plasmas of ADITYA-U. Further, ADITYA-U operations emphasize preventing disruptions and runaway electrons (REs) to ensure safe operations for future fusion devices. Significant suppression of REs has been achieved in ADITYA-U with the application of pulsed localized vertical magnetic field (LVF) perturbation, thereby establishing the technique’s independence from the tokamak device. The successful RE mitigation requires a critical threshold of LVF pulse magnitude, which is approximately 1% of the toroidal magnetic field, and a minimum duration of ∼5 ms. Apart from this, several novel findings have been achieved in the ADITYA-U experiments, including the modification of sawtooth duration through gas-puff, the emergence of MHD-induced geodesic acoustic mode-like oscillations, the propagation of fast heat pulses induced by MHD activity, the control of RE dynamics through Gas-puffs, the propagation of pinch-driven cold-pulses, the transport and core accumulations of argon impurities, the mass dependency of plasma toroidal rotation and the detection of ‘RICE’ scaling, as well as the characterization of edge plasma using wall conditioning methods, such as glow discharge cleaning using a combination of Ar -H _2 mixture, localized wall cleaning by electron cyclotron resonant plasma, and the development of machine learning-based disruption predictions, will be discussed in this paper.

Intrinsic toroidal rotation velocity ( V _φ ) has been measured from the Doppler shift of C ^5+ carbon spectral lines (at 529.05 nm) in the edge region of the ADITYA-U tokamak without any auxiliary torque input in an ohmically heated pure hydrogen (H _2 ) plasma as well as in H _2 plasmas seeded with medium-Z (neon and argon) impurities . The toroidal rotation in the edge region is observed to reverse its direction from the counter-current to the co-current direction with an increase in plasma current beyond I _p ∼ 145–150 kA. Furthermore, a systematic decrease in the co-current V _φ has been observed with the edge density, which tends to decrease to almost zero velocity with an increase in the edge density. The injection of medium- Z (neon and argon) impurities is observed to influence the edge toroidal rotation significantly. In low I _p discharges, argon injection leads to a reversal of edge intrinsic rotation from the counter-current to the co-current direction. In high I _p discharges, both neon and argon seeding enhance the co-current rotation by about ∼5–10 km s ^−1 , at a constant I _p compared to pure H _2 discharges. Simultaneous measurements of the edge radial electric field, E _r , shows that the E _r × B _θ flow seems to be driving the edge toroidal rotation in ADITYA-U. With impurity injection, the E _r also gets modified, leading to an observed increase in the edge toroidal rotation.

Effective control of impurities and precise regulation of the fueling gas are supreme prerequisites for optimal operation in any fusion device. Conventional wall-conditioning methods fall short of achieving optimal wall conditioning. Conventional wall-conditioning methods, such as vessel baking and H _2 /(D _2 )-fueled glow discharge cleaning (GDC), are generally required to remove wall-absorbed impurities in bulk after vessel venting. The excess amount of hydrogen, injected during H _2 GDC, can be reduced by helium (He)-fueled GDC. However, He removal from the vessel is more challenging due to its low molecular mass, very low condensation temperature, and inert characteristics. In ADITYA-U, optimal wall conditioning cannot be achieved using H _2 followed by He-fueled GDC when applied for extended periods spanning hours or days. A GDC with a mixture of argon and hydrogen (Ar–H _2 ) is introduced in the ADITYA-U tokamak to obtain better wall conditioning than H _2 followed by He GDC. In Ar–H _2 GDC, long-lived ArH ^+ ions are formed in sufficient numbers and accelerated toward the vessel wall with high momentum. This results in the breaking of high energy bonds of impurities with the wall/plasma facing components, which is not possible by H ^+ , H _2 ^+, H _3 ^+ ions in H _2 GDC due to their lower momentum. An optimal blend ratio of Ar to H _2 is established at 15%–20% for the mixture. This composition ensures that the introduction of high- Z Ar does not adversely affect tokamak plasma operations. The C- and O-containing impurities are reduced beyond the limit of the prolonged operation of H _2 GDC. Relative low pressures of dominant impurities such as CO, CH _4 , and H _2 O are obtained due to the Ar–H _2 GDC compared to routinely operated H _2 GDC. A comparison study of H _2 GDC and the developed Ar–H _2 GDC is performed in terms of wall conditioning and tokamak plasma operation. The encouraging results of the Ar–H _2 GDC are obtained in both wall cleaning and tokamak operation scenarios in the midsize tokamak ADITYA-U. This development and application of Ar–H _2 GDC are beneficial for large-sized fusion devices, leading to improved impurity reduction, reduced operational fuel consumption (H _2 /D _2 /He), and enhanced control over fuel recycling/extraction.

To reduce the risk of severe damage to the vessel and inner peripherals of any tokamak and its safe operation, a robust technique for the mitigation of runaway electrons (REs) is required. The REs in ADITYA tokamak are effectively mitigated by an application of local vertical magnetic field (LVF) perturbation. The LVF perturbation is applied using a pair of electromagnetic coils placed at the top and bottom of the ADITYA vacuum vessel in a Helmholtz configuration at one toroidal location. Powered by a capacitor bank power supply, these coils can produce a localized vertical magnetic field at the plasma center in the range of ∼150 G–260 G for a variable duration of 5–20 ms. The LVF pulse is first applied at the breakdown/current-ramp phase, where the REs are generated in the discharges initiated by the conventional ohmic breakdown in ADITYA. With the application of LVF pulse the REs are significantly reduced as indicated by the reduction in the REs generated hard x-ray flux. It has been observed that to extract the REs efficiently, an LVF pulse of magnitude at least ∼1% of the toroidal magnetic field with a minimum duration of ∼5 ms should be applied. The LVF perturbation is applied at different times into the discharge, i.e. during the breakdown/current ramp-up phase and current flat-top phase. The REs are significantly reduced in all the phases and improved discharge consistency. The LVF acts as an error field and a short-pulse of the LVF influences the REs more in comparison to the thermal electrons due to the faster velocities of the REs.

The effect of a periodic train of short gas-puff pulses on the rotation frequency and amplitude of drift-tearing modes has been studied in ADITYA/ADITYA-U tokamak. The short gas puffs, injecting approximately molecules of fuel gas (hydrogen) at one toroidal location, are found to concomitantly decrease the drift-tearing mode rotation frequency and the mode amplitude during the period of injection and then recover back to its initial values when the gas pulse is over. This leads to a periodic modulation of the rotation frequency and amplitude of the drift-tearing modes that is correlated with the periodicity of the gas pulse injection. The underlying mechanism for this change in the mode characteristic appears to be related to gas puff induced change in the radial profile of the plasma pressure in the edge region that brings about a reduction in the diamagnetic drift frequency. Detailed experimental measurements and BOUT++ code simulations support such a reduction in diamagnetic drift frequency. Our results reveal a close interaction between the edge dynamics and core MHD phenomena in a tokamak that could help us better understand the rotation dynamics and amplitude pulsations of magnetic islands. [ABSTRACT FROM AUTHOR]

Two dimensional interchange turbulence has been used to describe neon gas interaction with tokamak plasma, and is able to include anomalous transport effects self-consistently in the edge and scrape-off layer (SOL) regions. Model equations related to the plasma turbulence coupled with neon gas have been solved numerically using the BOUT++ framework code. Two different fluid models of the neon gas have been investigated, and in this context a few results related to the Aditya tokamak are presented. Numerical results indicate that ionization and radiative cooling processes modify the plasma turbulence. The existence of neon ions and their radial profile in the edge and SOL regions are the most significant results in this work. This new result has been investigated in detail in the presence of the interchange plasma turbulence. Other relevant results such as an increase of plasma density and a small change of electron temperature are also presented. In turbulence saturated states the electron temperature can increase slightly even in the presence of radiative cooling. Neon ions in the presence of polarization drift and radiative cooling modify the radial electric field and its radial shear. Radial electron energy flux has been investigated in the context of poloidal velocity shear and turbulence decorrelation rates. An increase of the electron energy confinement time has been observed numerically, mainly in the SOL region in the presence of gas seeding. Experimental results show an increase of global energy confinement time only when the gas seeding becomes effective in changing the global plasma parameters. [ABSTRACT FROM AUTHOR]

The first Indian tokamak, ADITYA, operated for over two decades with a circular poloidal limiter, has been upgraded to a tokamak named ADITYA Upgrade (ADITYA-U) to attain shaped-plasma operations with an open divertor in single and double-null configurations. Experimental research using ADITYA-U has made significant progress since the last FEC in 2016. After installation of a plasma facing component and standard tokamak diagnostics in ADITYA-U, the Phase-I plasma operations were initiated in December 2016 with a graphite toroidal belt limiter. Ohmically heated circular plasmas supported by filament pre-ionization with plasma parameters Ip ~ 80–95 kA, duration ~80–180 ms, with a maximum toroidal field ~1 T and chord averaged electron density ~2.5  ×  1019 m−3, have been obtained. The runaway electron (RE) generation, transport and mitigation experiments, along with magneto hydrodynamic (MHD) activities and density enhancement with H2 gas puffing experiments were carried out in Phase-I, which was completed in March 2017. Preparation for the Phase-II operation in ADITYA-U includes calibration of magnetic diagnostics followed by commissioning of major diagnostics and installation of a baking system. After repeated cycles of baking the vacuum vessel up to ~135 °C, the Phase-II operations resumed in February 2018 and are continuing to achieve plasma parameters close to the design parameters of circular limiter plasmas, using real-time plasma position control. The plasma current has been raised to ~135 kA in Phase-II, with a maximum chord averaged electron density of ~4  ×  1019 m−3. Hydrogen gas breakdown has been observed in more than 2000 discharges, including Phase-I and Phase-II operations, without a single failure. Several experiments have been carried out, including the control of REs with the fuelling of supersonic molecular beam injection as well as sonic H2 gas puffing during current flat-top, MHD mode studies using multiple periodic gas puffs, and radiative improved modes using neon gas puffs. The experimental results from Phase-I and Phase-II operations of the ADITYA-U tokamak are discussed in this paper. [ABSTRACT FROM AUTHOR]

An impurity ion toroidal rotation profile has been observed in the ohmic discharges of the Aditya-U tokamak from the Doppler-shifted passive charge exchange line emission of C5+ at 529 nm. The line has been monitored using a space-resolved visible spectroscopy diagnostic consisting of a 1 m multi-track spectrometer coupled with a charge coupled device (CCD) detector. In typical discharges with central chord-averaged electron density  ≲2.5  ×  1019 m−3, the maximum toroidal rotation velocity at the plasma core is found to be  −20 km s−1 in the counter-current direction. Reversal of plasma toroidal rotation is observed for high electron density discharges with a central chord-averaged density of ~5  ×  1019 m−3. During the rotation reversal, the direction of toroidal rotation changes to the co-current direction in the core. Interestingly, in the discharges with a central chord-averaged electron density of  ≲2.5  ×  1019 m−3, where neon gas is injected at the plasma current flat-top, the toroidal rotation reversal has been observed after the gas puff. [ABSTRACT FROM AUTHOR]

The neutral particle temperature in the edge region of the Aditya-U tokamak has been measured by recording the hydrogen Balmer alpha emission spectra at 656.28 nm from different lines of sight in both the high and low field sides. The spatial profile of the Hα emission has been recorded using a 1 m multi-track spectrometer. The neutral temperature is estimated from the Doppler broadening of the measured Hα spectrum by appropriately removing the contribution from the Zeeman splitting of the spectral lines. A computer simulation code was developed to estimate the neutral temperature by including the broadening mechanisms such as Doppler broadening and the Zeeman effect to simulate the Hα emission spectrum along with the proper convolution of the instrumental width of the diagnostic system. It has been observed that the high field side neutral temperature (~4–6 eV) during plasma flat top is almost twice that of the low field side (~2–3 eV) suggesting poloidal asymmetry in the neutral temperature. [ABSTRACT FROM AUTHOR]

In the ADITYA Upgrade tokamak, glow discharge wall conditioning (GDC) is performed regularly during the high-temperature plasma operation cycle using hydrogen (H) and helium (He) gases. H GDC is carried out after long durations (few hours) of plasma operations on every plasma operation day in automatic mode to control oxygen (O) and carbon (C) containing impurities. This leads to high retention of H gas on graphite limiter plates and stainless steel (SS) vessel walls. Subsequently, the high outgassing of H requires a prolonged pumping time and high H recycling during plasma discharges affects the plasma performance in respect to H fueling control of the plasma. To overcome the above-mentioned issues with continuous H GDC for longer durations, a new approach involving pulsed glow discharge wall conditioning (P-GDC) has been introduced in the ADITYA-U tokamak to reduce the residual H and He concentration in SS vessel walls and graphite limiter plates. To facilitate the fast initiation of a discharge in the case of P-GDC, a source of free electrons from a hot filament has been introduced in the vessel. A fast feedback controlled pulsed-gas-fueling system has been developed to initiate a glow discharge in each gas-feed pulse at various operating pressures from 1  ×  10−4 Torr to 10−3 Torr in the presence of an applied DC voltage. The different P-GDC experiments have been carried out with H, He and argon as the working gases and the results are compared with traditional continuous GDC. The P-GDC experiments have been optimized to provide beneficial wall conditioning for plasma operations. In this paper, the design, development and operation of P-GDC has been described along with the preliminary studies of its effect on the measured impurity line radiation during a plasma discharge. [ABSTRACT FROM AUTHOR]

Several experiments, related to controlled thermonuclear fusion research and highly relevant for large size tokamaks, including ITER, have been carried out in ADITYA, an ohmically heated circular limiter tokamak. Repeatable plasma discharges of a maximum plasma current of ~160 kA and discharge duration beyond ~250 ms with a plasma current flattop duration of ~140 ms have been obtained for the first time in ADITYA. The reproducibility of the discharge reproducibility has been improved considerably with lithium wall conditioning, and improved plasma discharges are obtained by precisely controlling the position of the plasma. In these discharges, chord-averaged electron density ~3.0–4.0  ×  1019 m−3 using multiple hydrogen gas puffs, with a temperature of the order of ~500–700 eV, have been achieved. Novel experiments related to disruption control are carried out and disruptions, induced by hydrogen gas puffing, are successfully mitigated using the biased electrode and ion cyclotron resonance pulse techniques. Runaway electrons are successfully mitigated by applying a short local vertical field (LVF) pulse. A thorough disruption database has been generated by identifying the different categories of disruption. Detailed analysis of several hundred disrupted discharges showed that the current quench time is inversely proportional to the q edge. Apart from this, for volt–sec recovery during the plasma formation phase, low loop voltage start-up and current ramp-up experiments have been carried out using electron cyclotron resonance heating (ECRH). Successful recovery of volt–sec leads to the achievement of longer plasma discharge durations. In addition, the neon gas puff assisted radiative improved confinement mode has also been achieved in ADITYA. All of the above mentioned experiments will be discussed in this paper. [ABSTRACT FROM AUTHOR]

This paper summarizes the results of recent dedicated experiments on disruption control and runaway mitigation carried out in ADITYA, which are of the utmost importance for the successful operation of large size tokamaks, such as ITER. It is quite a well-known fact that disruptions in tokamaks must be avoided. Disruptions, induced by hydrogen gas puffing, are successfully avoided by two innovative techniques in ADITYA using a bias electrode placed inside the last closed flux surface and applying an ion cyclotron resonance pulse with a power of ∼50 to 70 kW. These experiments led to better understanding of the disruption avoidance mechanisms and also can be thought of as one of the options for disruption avoidance in ITER. In both cases, the physical mechanism seems to be the control of magnetohydrodynamic modes due to increased poloidal rotation of edge plasma generated by induced radial electric fields. Real time avoidance of disruption with identifying proper precursors in both the mechanisms is successfully attempted. Further, analysing thoroughly the huge database of different types of spontaneous and deliberately-triggered disruptions from ADITYA, a significant contribution has been made to the international disruption database (ITPA). Furthermore, the mitigation of the runaway electron generated mainly during disruptions remains a challenging topic in present tokamak research as these high-energy electrons can cause severe damage to in-vessel components and the vacuum vessel. A simple technique has been implemented in ADITYA to mitigate the runaway electrons before they can gain high energies using a localized vertical magnetic field perturbation applied at one toroidal location to extract runaway electrons. [ABSTRACT FROM AUTHOR]